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Category:CORRESPONDENCE-LETTERS
MONTHYEARML18039A9021999-10-15015 October 1999 Forwards LER 99-010-00 Re Occurrence of Plant Reactor Scram Due to Main Turbine Trip Which Resulted in Main Steam Moisture Separator.All Plant Safety Systems Operated as Designed in Response to Event ML20217E0711999-10-14014 October 1999 Grants Approval for Util to Submit Original,One Signed Paper Copy & Six CD-ROM Copies of Updates to FSAR as Listed,Per 10CFR50.4(c),in Response to ML18039A8961999-10-14014 October 1999 Forwards LER 99-009-00,re Manual Reactor Scram on Unit 2 from 54% Power,Iaw 10CFR50.73(a)(2)(iv).All Plant Safety Sys Operated as Designed in Response to Event ML20217D3261999-10-0808 October 1999 Responds to Re Event Concerning Spent Fuel Pool Water Temperature Being Undetected for Approx Two Days at Browns Ferry Unit 3 ML20217F7751999-10-0808 October 1999 Confirms 991006 Telcon Between T Abney of Licensee Staff & a Belisle of NRC Re Meeting to Be Conducted on 991109 in Atlanta,Ga to Discuss Various Maintenance Issues ML18039A8931999-10-0808 October 1999 Forwards LER 99-008-00,concerning HPCI Sys Being Declared Inoperable,Iaw 10CFR50.73(a)(2)(v).There Are No Commitments Contained in Ltr ML18039A8881999-10-0808 October 1999 Provides Licensee Supplemental Response to NRC 980713 RAI Re GL 87-02,Suppl 1, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors. ML20217B5481999-10-0101 October 1999 Requests Exception to 10CFR50.4(c) Requirement to Provide Total of Twelve Paper Copies When Submitting Revs to BFN UFSAR ML20212M1481999-09-28028 September 1999 Refers to Management Meeting Conducted on 990927 at Region II for Presentation of Recent Plant Performance.List of Attendees & Copy of Presentation Handout Encl ML20212F7751999-09-22022 September 1999 Requests Operator & Senior Operator License Renewals for Listed Individuals and Licenses ML20212D3651999-09-20020 September 1999 Forwards SE Accepting Licensee 990430 Proposed Rev to Plant, Unit 3 Matl Surveillance Program ML18039A8721999-09-10010 September 1999 Informs of Licensee Decision to Withdraw Proposed Plant risk-informed Inservice Insp Program,Originally Transmitted in Util 981023 Ltr.Licensee Expects to Resubmit Revised Program within Approx 6 Wks ML20211Q5731999-09-0909 September 1999 Submits Response to Administrative Ltr 99-03 Re Preparation & Scheduling of Operator Licensing Exams.Completed NRC Form 536,operator Licensing Exam Data,Which Provides Plant Current Schedules for Specific Info Requested Encl ML20211G6491999-08-26026 August 1999 Confirms Telcon with T Abney on 990824 Re Mgt Meeting Which Has Been re-scheduled from 990830-0927.Purpose of Meeting to Discuss BFN Status & Performance ML20210Q6931999-08-0909 August 1999 Forwards Updated Changes to Distribution Lists for Browns Ferry & Bellefonte Nuclear Plants ML18039A8371999-08-0606 August 1999 Forwards BFN Unit 2 Cycle 10 ASME Section XI NIS-1 & NIS-2 Data Repts, for NRC Review.Corrected Inservice Insp Summary Rept for Unit 3 Cycle 8 Operation,Included in Rept ML20210Q4421999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006. Authorized Representative of Facility Must Submit Ltr with List of Individuals to Take exam,30 Days Before Exam Date ML20210N1051999-08-0202 August 1999 Forwards SE Accepting Licensee 990326 Request for Relief from ASME B&PV Code,Section XI Requirements.Request for Relief 3-ISI-7,pertains to Second 10-year Interval ISI for Plant,Unit 3 ML20210G8991999-07-28028 July 1999 Discusses 990726 Open Mgt Meeting for Discussion on Plant Engineering Status & Performance.List of Attendees & Presentation Handout Encl ML18039A8181999-07-26026 July 1999 Forwards LER 99-004-00 Re Inoperability of Two Divisions of Plant CSS Due to Personnel Error During Surveillance Testing.Event Reported Per 10CFR50.73(a)(2)(i)(B) ML20210G8051999-07-22022 July 1999 Discusses DOL Case DC Smith Vs TVA Investigation.Oi Concluded That There Was Not Sufficient Evidence Developed During Investigation to Substantiate Discrimination.Nrc Providing Results of OI Investigation to Parties ML20210F3031999-07-22022 July 1999 Submits Rept Re Impact of Changes or Errors in Methodology Used to Demonstrate Compliance with ECCS Requirements of 10CFR50.46.One Reportable non-significant Error Was Found During Time Period of 980601-990630 ML20209J0251999-07-16016 July 1999 Forwards SE Which Constitutes Staff Review & Approval of TVA Ampacity Derating Test & Analyses for Thermo-Lag Fire Barrier Configurations as Required in App K of Draft Temporary Instruction, Fpfi, ML20210B2671999-07-14014 July 1999 Confirms 990702 Telcon Between T Abney of Licensee Staff & Author Re Mgt Meeting Scheduled for 990830 at Licensee Request in Atlanta,Ga to Discuss Browns Ferry Nuclear Plant Status & Performance ML20209E3421999-07-0707 July 1999 Confirms Arrangements Made During 990628 Telephone Conversation to Hold Meeting on 990726 in Atlanta,Ga to Discuss Plant Engineering Status & Performance ML20209E5511999-07-0707 July 1999 Informs That as Result of NRC Review of Util Responses to GL 92-01,rev 1 & Suppl 1 & Suppl 1 Rai,Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2.This Closes TACs MA1180,MA1181 & MA1179 ML20196J3531999-06-30030 June 1999 Responds to Re Boeing Rocket Booster Mfg Facility Being Constructed in Decatur,Al.Nrc Has No Unique Emergency Planning Concerns Re Proximity of Boeing Facility to BFN ML20196G9111999-06-28028 June 1999 Discusses Completion of Licensing Action for GL 96-01, Testing of Safety-Related Logic Circuits ML18039A8081999-06-28028 June 1999 Forwards LER 99-004-00 Re Esfas That Occurred When RPS Motor Generator Tripped.Rept Is Submitted IAW Provisions of 10CFR50.73(a)(2)(iv) as Event of Condition That Resulted in Automatic Actuation of ESF ML18039A8111999-06-25025 June 1999 Requests Permanent Relief from Inservice Insp Requirements of 10CFR50.55a(g) for Volumetric Exam of Bfn,Unit 3 Circumferential RPV Welds,Per GL 98-05 ML20196F8741999-06-23023 June 1999 Forwards Safety Evaluation Accepting Util Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20196F8131999-06-22022 June 1999 Forwards Rev 24 to Security Personnel Training & Qualification Plan,Per 10CFR50.54(p).Rev Withheld ML18039A8051999-06-22022 June 1999 Forwards LER 99-003-00,re Automatic Reactor Scram Due to Turbine Trip.Rept Numbered 99-001 Should Be Deleted & Replaced with Encl Rept as Result of Error Noted in 990614 Rept ML18039A8031999-06-18018 June 1999 Responds to NRC Staff Verbal Request Re TS Change TS-376, Originally Submitted on 970312, & Proposed Changes to TS to Extend Current 7-day AOT for EDGs to 14 Days ML18039A7931999-06-0101 June 1999 Provides Summary of Major Activities Performed at BFN During Scheduled Unit 2 Cycle 10 Refueling Outage ML20195D3321999-06-0101 June 1999 Informs That Cb Fisher,License OP-5525-4,can No Longer Maintain License at Plant Because of Physical Condition That Causes Licensee to Fail to Meet Requirements of 10CFR55.21 ML18039A7911999-05-24024 May 1999 Informs That by Meeting Test Criteria Established by Test Based on Ansi/Ans 3.5-1985 (License Amends 254 & 214) power- Uprate Simulation Acceptable for Operator Training ML18039A7891999-05-24024 May 1999 Informs That Oscillation Power Range Monitor Module Has Been Enabled for Current Cycle of Operation Following Unit 2 Cycle 10 Refueling Outage Which Was Completed on 990509 ML20195B9361999-05-24024 May 1999 Informs That Do Elkins,License SOP-3392-6,no Longer Needs to Maintain License as Position Does Not Require License ML20206U6551999-05-14014 May 1999 Informs That ML Meek & Wd Dawson Will No Longer Need to Maintain SRO Licenses at Plant,Due to Termination of Employment,Effective 990521 ML20206Q8421999-05-10010 May 1999 Forwards Medical Info on DM Olive,License SOP-20540-2,in Response to NRC 990428 Telcon.Encl Withheld from Public Disclosure IAW 10CFR2.790(a)(6) ML18039A7771999-05-0606 May 1999 Forwards LER 99-003-00,providing Details Re Plant HPCI Sys Being Declared Inoperable Due to Loose Electrical Connection.Ltr Contains No Commitments ML20206G6611999-05-0404 May 1999 Forwards SE Accepting GL 88-20,submitted by TVA Re multi-unit Probabilistic Risk Assessement (Mupra) for Plant, Units 1,2 & 3 ML18039A7741999-04-30030 April 1999 Forwards Proposed Rev to BFN Unit 3 RPV Matl Surveillance Program,For NRC Approval ML20206H5901999-04-30030 April 1999 Forwards Notification of Revs to BFN Unit 2 Emergency Response Data Sys Data Point Library.Revs Were Implemented on 990413 DD-99-06, Informs That Time Provided by NRC within Which Commission May Act to Review Director'S Decision (DD-99-06) Has Expired.Decision Became Final Agency Action on 990423.With Certificate of Svc.Served on 9904281999-04-28028 April 1999 Informs That Time Provided by NRC within Which Commission May Act to Review Director'S Decision (DD-99-06) Has Expired.Decision Became Final Agency Action on 990423.With Certificate of Svc.Served on 990428 ML18039A7681999-04-27027 April 1999 Requests Relief from Specified Inservice Insp Requirements in Section XI of ASME Boiler & Pressure Vessel Code,Per 10CFR50.55a(a)(3)(i).Relief Requests 2-ISI-8 & 3-ISI-8,encl for NRC Review & Approval ML18039A7591999-04-27027 April 1999 Forwards Annual Radiological Environ Operating Rept Browns Ferry Nuclear Plant 1998. Rept Includes Results of Land Use Censuses,Summarized & Tabulated Results of Radiological Environ Samples in Format of Reg Guide 4.8 & NUREG-1302 ML18039A7651999-04-27027 April 1999 Forwards Rev 0 to TVA-COLR-BF2C11, Browns Ferry Nuclear Plant Unit 2,Cycle 11 Colr. ML18039A7541999-04-23023 April 1999 Requests Approval of Bfnp Unit 3 Risk-Informed ISI (RI-ISI) Program,Per 10CFR50.55(a)(3)(i) & GL 88-01.Encl RI-ISI Program Is Alternative to Current ASME Section XI ISI Requirments for Code Class 1,2 & 3 Piping 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML18039A9021999-10-15015 October 1999 Forwards LER 99-010-00 Re Occurrence of Plant Reactor Scram Due to Main Turbine Trip Which Resulted in Main Steam Moisture Separator.All Plant Safety Systems Operated as Designed in Response to Event ML18039A8961999-10-14014 October 1999 Forwards LER 99-009-00,re Manual Reactor Scram on Unit 2 from 54% Power,Iaw 10CFR50.73(a)(2)(iv).All Plant Safety Sys Operated as Designed in Response to Event ML18039A8881999-10-0808 October 1999 Provides Licensee Supplemental Response to NRC 980713 RAI Re GL 87-02,Suppl 1, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors. ML18039A8931999-10-0808 October 1999 Forwards LER 99-008-00,concerning HPCI Sys Being Declared Inoperable,Iaw 10CFR50.73(a)(2)(v).There Are No Commitments Contained in Ltr ML20217B5481999-10-0101 October 1999 Requests Exception to 10CFR50.4(c) Requirement to Provide Total of Twelve Paper Copies When Submitting Revs to BFN UFSAR ML20212F7751999-09-22022 September 1999 Requests Operator & Senior Operator License Renewals for Listed Individuals and Licenses ML18039A8721999-09-10010 September 1999 Informs of Licensee Decision to Withdraw Proposed Plant risk-informed Inservice Insp Program,Originally Transmitted in Util 981023 Ltr.Licensee Expects to Resubmit Revised Program within Approx 6 Wks ML20211Q5731999-09-0909 September 1999 Submits Response to Administrative Ltr 99-03 Re Preparation & Scheduling of Operator Licensing Exams.Completed NRC Form 536,operator Licensing Exam Data,Which Provides Plant Current Schedules for Specific Info Requested Encl ML20210Q6931999-08-0909 August 1999 Forwards Updated Changes to Distribution Lists for Browns Ferry & Bellefonte Nuclear Plants ML18039A8371999-08-0606 August 1999 Forwards BFN Unit 2 Cycle 10 ASME Section XI NIS-1 & NIS-2 Data Repts, for NRC Review.Corrected Inservice Insp Summary Rept for Unit 3 Cycle 8 Operation,Included in Rept ML18039A8181999-07-26026 July 1999 Forwards LER 99-004-00 Re Inoperability of Two Divisions of Plant CSS Due to Personnel Error During Surveillance Testing.Event Reported Per 10CFR50.73(a)(2)(i)(B) ML20210F3031999-07-22022 July 1999 Submits Rept Re Impact of Changes or Errors in Methodology Used to Demonstrate Compliance with ECCS Requirements of 10CFR50.46.One Reportable non-significant Error Was Found During Time Period of 980601-990630 ML18039A8081999-06-28028 June 1999 Forwards LER 99-004-00 Re Esfas That Occurred When RPS Motor Generator Tripped.Rept Is Submitted IAW Provisions of 10CFR50.73(a)(2)(iv) as Event of Condition That Resulted in Automatic Actuation of ESF ML18039A8111999-06-25025 June 1999 Requests Permanent Relief from Inservice Insp Requirements of 10CFR50.55a(g) for Volumetric Exam of Bfn,Unit 3 Circumferential RPV Welds,Per GL 98-05 ML20196F8131999-06-22022 June 1999 Forwards Rev 24 to Security Personnel Training & Qualification Plan,Per 10CFR50.54(p).Rev Withheld ML18039A8051999-06-22022 June 1999 Forwards LER 99-003-00,re Automatic Reactor Scram Due to Turbine Trip.Rept Numbered 99-001 Should Be Deleted & Replaced with Encl Rept as Result of Error Noted in 990614 Rept ML18039A8031999-06-18018 June 1999 Responds to NRC Staff Verbal Request Re TS Change TS-376, Originally Submitted on 970312, & Proposed Changes to TS to Extend Current 7-day AOT for EDGs to 14 Days ML18039A7931999-06-0101 June 1999 Provides Summary of Major Activities Performed at BFN During Scheduled Unit 2 Cycle 10 Refueling Outage ML20195D3321999-06-0101 June 1999 Informs That Cb Fisher,License OP-5525-4,can No Longer Maintain License at Plant Because of Physical Condition That Causes Licensee to Fail to Meet Requirements of 10CFR55.21 ML20195B9361999-05-24024 May 1999 Informs That Do Elkins,License SOP-3392-6,no Longer Needs to Maintain License as Position Does Not Require License ML18039A7911999-05-24024 May 1999 Informs That by Meeting Test Criteria Established by Test Based on Ansi/Ans 3.5-1985 (License Amends 254 & 214) power- Uprate Simulation Acceptable for Operator Training ML18039A7891999-05-24024 May 1999 Informs That Oscillation Power Range Monitor Module Has Been Enabled for Current Cycle of Operation Following Unit 2 Cycle 10 Refueling Outage Which Was Completed on 990509 ML20206U6551999-05-14014 May 1999 Informs That ML Meek & Wd Dawson Will No Longer Need to Maintain SRO Licenses at Plant,Due to Termination of Employment,Effective 990521 ML20206Q8421999-05-10010 May 1999 Forwards Medical Info on DM Olive,License SOP-20540-2,in Response to NRC 990428 Telcon.Encl Withheld from Public Disclosure IAW 10CFR2.790(a)(6) ML18039A7771999-05-0606 May 1999 Forwards LER 99-003-00,providing Details Re Plant HPCI Sys Being Declared Inoperable Due to Loose Electrical Connection.Ltr Contains No Commitments ML20206H5901999-04-30030 April 1999 Forwards Notification of Revs to BFN Unit 2 Emergency Response Data Sys Data Point Library.Revs Were Implemented on 990413 ML18039A7741999-04-30030 April 1999 Forwards Proposed Rev to BFN Unit 3 RPV Matl Surveillance Program,For NRC Approval ML18039A7681999-04-27027 April 1999 Requests Relief from Specified Inservice Insp Requirements in Section XI of ASME Boiler & Pressure Vessel Code,Per 10CFR50.55a(a)(3)(i).Relief Requests 2-ISI-8 & 3-ISI-8,encl for NRC Review & Approval ML18039A7591999-04-27027 April 1999 Forwards Annual Radiological Environ Operating Rept Browns Ferry Nuclear Plant 1998. Rept Includes Results of Land Use Censuses,Summarized & Tabulated Results of Radiological Environ Samples in Format of Reg Guide 4.8 & NUREG-1302 ML18039A7651999-04-27027 April 1999 Forwards Rev 0 to TVA-COLR-BF2C11, Browns Ferry Nuclear Plant Unit 2,Cycle 11 Colr. ML20206C8591999-04-23023 April 1999 Informs That Util Has Determined,Dr Bateman No Longer Needs to Maintain His License,Effective 990331,per Requirement of 10CFR55.55(a) ML18039A7541999-04-23023 April 1999 Requests Approval of Bfnp Unit 3 Risk-Informed ISI (RI-ISI) Program,Per 10CFR50.55(a)(3)(i) & GL 88-01.Encl RI-ISI Program Is Alternative to Current ASME Section XI ISI Requirments for Code Class 1,2 & 3 Piping ML18039A7581999-04-23023 April 1999 Responds to Item 4 of 981117 RAI Re TS Change Request 376 Re Extended EDG Allowed Outage Time,In Manner Consistent with Rgs 1.174 & 1.177 ML20206C1241999-04-21021 April 1999 Forwards Annual Occupational Radiation Exposure Rept for 1998, IAW TS Section 5.6.1.Rept Reflects Radiation Exposure Data as Tracked by Electronic Dosimeters on Radiation Work Permits ML20205T0971999-04-15015 April 1999 Submits Change in Medical Status for DM Olive in Accordance with 10CFR55.25,effective 990315.Encl Medical Info & Certification of Medical Exam,Considered by Util to Be of Personal Nature & to Be Withheld,Per 10CFR2.790(a)(6) ML18039A7441999-04-0707 April 1999 Forwards LER 99-001-00,providing Details Re Inoperability of Two Trains of Standby Gas Treatment Due to Breaker Trip on One Train in Conjunction with Planned Maint Activities on Other.Ltr Contains No New Commitments ML18039A7431999-03-30030 March 1999 Responds to NRC 990112 RAI Re BFN Program,Per GL 96-05, Periodic Verification of Design-Basis Capability of Safety- Related Movs. ML18039A7421999-03-30030 March 1999 Provides Results of Analysis of Design Basis Loca,As Required by License Condition Re Plants Power Uprate Operating License Amends 254 & 214 ML18039A7411999-03-30030 March 1999 Provides Partial Response to NRC 981117 RAI Re TS Change Request 376,proposing to Extend Current 7 Day AOT for EDG to 14 Days ML18039A7371999-03-26026 March 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME Boiler & Pressure Vessel Code,1989 Edition.Encl Contains Request for Relief 3-ISI-7,for NRC Review & Approval ML18039A7331999-03-26026 March 1999 Forwards Rev 4 to TVA-COLR-BF2C10, Bnfp,Unit 2,Cycle 10 COLR, IAW Requirements of TS 5.6.5.d.COLR Was Revised to Extend Max Allowable Nodal Exposure for GE GE7B Fuel Bundles ML18039A7291999-03-22022 March 1999 Forwards Revised Epips,Including Index,Rev 26A to EPIP-1, Emergency Classification Procedure & Rev 26A to EPIP-5, General Emergency. Rev 26A Includes All Changes Made in Rev 26 as Well as Identified Errors ML20204G8471999-03-19019 March 1999 Reports Change in Medical Status for Ma Morrow,In Accordance with 10CFR55.25.Encl Medical Info & Certification of Medical Exam,Considered by Util to Be of Personal Nature & to Be Withheld from Pdr,Per 10CFR2.790(a)(6).Without Encl ML20207M0611999-03-11011 March 1999 Forwards Goals & Objectives for May 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3,radiological Emergency Plan Exercise.Plant Exercise Is Currently Scheduled for Wk of 990524 ML18039A6971999-02-22022 February 1999 Forwards Typed TS Pages,Reflecting NRC Approved TS Change 354 Requiring Oscillation PRM to Be Integrated Into Approved Power uprate,24-month Operating Cycle & Single Recirculation Loop Operation ML18039A6961999-02-19019 February 1999 Provides Util Response to GL 95-07 Re RCIC Sys Injection Valves (2/3-FCV-71-39) for BFN Units 2 & 3.Previous Responses,Dtd 951215,1016 & 960730,0315 & 0213,supplemented ML18039A6911999-02-19019 February 1999 Forwards Rev 3 to Unit 2 Cycle 10 & Rev 1 to Unit 3 Cycle 9, Colr.Colrs for Each Unit Were Revised to Include OLs Consistent with Single Recirculation Loop Operation ML20203B6031999-02-0404 February 1999 Requests Temporary Partial Exemption from Requirements of 10CFR50.65,maint Rule for Unit 1.Util Requesting Exemption to Resolve Issue Initially Raised in NRC Insp Repts 50-259/97-04,50-260/97-04 & 50-296/97-04,dtd 970521 ML18039A6741999-01-21021 January 1999 Responds to NRC 981209 Ltr Re Violations Noted in Insp Repts 50-259/98-07,50-260/98-07 & 50-296/98-07,respectively. Corrective Actions:Will Revise Procedure NEPD-8 Re Vendor Nonconformance Documentation Submission to TVA ML20199F6951999-01-0808 January 1999 Submits Request for Relief from ASME Section XI Inservice Testing Valve Program to Extend Interval Between Disassembly of Check Valve,Within Group of Four Similar Check Valves for EECW Dgs,From 18 to 24 Months 1999-09-09
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CATEGORY 1 REGULA'I Y INFORMATION DISTRIBUTIOI SYSTEM (RIDS)ACCESSION NBR:9806290033 DOC.DATE: 98/06/17 NOTARIZED:
NO DOCKET¹FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-,296 browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH.NAME AUTHOR AFFILIATION ABNEY,T.E.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Submits response to NRC 980603 RAI re Browns Ferry Nuclear Plant 971001 proposed TS change for power uprate operation.
DISTRIBUTION CODE: D030D COPIES RECEIVED:LTR
)ENCL I SIZE: JO TITLE: TVA Facilities
-Routine Correspondence NOTES: RECIPIENT ID CODE/NAME PD2-3 DEAGAZIO,A COPIES RECIPIENT LTTR ENCL ID CODE/NAME 1 1 v PD2-3-PD 1 1 COPIES LTTR ENCL 1 1 E INTERNAL: ACRS OGC/HDS3 EXTERNAL: NOAC 1 1 1'1 1 ILE CENT~ER 03.RES'/'DRESS EB/SES NRC PDR 1 1 1 1 1 1 D E NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 9 ENCL 8 A I'I C 1, l~'" d I Tennessee Valley Authority, Post Of(ice Box 2000, Decatur.Alabama 35609 June 17, 1998 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.20555 Gentlemen:
In the Matter of)Tennessee Valley Authority)Docket Nos.50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)-RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)RELATING TO UNITS 2 AND 3 TECHNICAL SPECIFICATION (TS)CHANGE NO.TS-384-POWER UPRATE OPERATION (TAC NOS.M99711 AND M99712)This letter provides additional information requested by NRC in support of TS-384.On October 1, 1997, TVA provided TS-384, an amendment to Operating Licenses DPR-52 and DPR-68 that will allow Units 2 and 3 to operate at an uprated power level of 3458 MWt.The enclosure provides TVA's response to the June 3, 1998, NRC RAI for the October 1, 1997, proposed TS change.This letter includes replies to each of the NRC requests.There are no new commitments made in this letter.If you have any questions, please telephone me at (256)729-2636.cerely,~~n Manager of Lice and Indust Affa'"t'$~O~QUSi.VV cc: See Pa e 3'sf806290033
'tf806'17 PDR'ADQCK 05000260 p PDR U.S.Nuclear Regulatory Commission Page 2'une"17, 1998 RE FE RENCE S 2.3.TVA letter to NRC dated October 1, 1997, Browns Ferry Nuclear Plant (BFN)-Units 2 and 3-Technical Specification (TS)Change TS-384-Request For License Amendment for Power Uprate Operation TVA letter to NRC dated March 16, 1998, Browns Ferry Nuclear Plant (BFN)-Units 2 and 3 Technical Specification (TS)No.384 Supplement 1-Request for License Amendment for Power Uprate Operation NRC letter to TVA dated June 3, 1998, Browns Ferry Nuclear Plant, Units 2 and 3: Request for Additional Information Relating to Technical Specification Change No.TS-384-Power Uprate Operation (TAC Nos.M99711 and M99712)
U.S.Nuclear Regulatory Commission Page 3'une 17, 1998 Enclosure cc (Enclosure):
Albert W.De Agazio, Project Manager U.S.Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr.Harold 0.Christensen, Branch Chief U.S.Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector BFN Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 L.Raghaven, Project Manager U.S.Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 BROWNS FERRY NUCLEAR PLANT (BFN)-RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)RELATING TO UNITS 2 AND 3 TECHNICAL SPECIFICATION (TS)CHANGE NO.TS-384-POWER UPRATE OPERATION (TAC NOS.M99711 AND M99712)This enclosure provides the TVA response to the June 3, 1998, NRC request for additional information.
NRC REQUEST 1 In Section 4.1.1.1, Containment long term pool temperature response, it is indicated that a pre-uprate SHEX code benchmark evaluation was performed using a heat exchanger K-factor of 223 BTU/sec-'F, 92'Residual heat removal (RHR)service water (SW)temperature, 6500 gpm RHR flow rate, and 4000 gpm RHR SW flow rate.The analysis resulted in a peak suppression pool temperature of 1750'[sic].These values are not consistent with the existing BFN analysis which used a heat exchanger K-factor of 228 BTU/sec-'F, 95'RHR service water temperature, 6500 gpm RHR flow rate, and 4500 gpm RHR SW flow rate and reported a peak pool temperature of 1770'[sic].The existing analysis was performed using the May-Witt decay heat curve while the uprate was performed using the ANS/ANSI 5.1 (2a uncertainty).
The comparative analysis should use the same input assumptions.
Please justify this inconsistency discrepancy.
Also, please provide the containment temperature and pressure profiles.TVA REPLY 1 A plant-specific SHEX benchmark case using inputs consistent with the Updated Final Safety Analysis Report (UFSAR)basis (i.e., heat exchanger K-factor of 228 BTU/sec-'F, RHR service water temperature of 95', 6500 gpm RHR flow rate, 4500 gpm RHR SW flow rate and May-Witt decay heat curve)was performed for BFN as part of the power uprate containment analyses.The resulting maximum suppression pool temperature is 176.7'.This result is in agreement with the UFSAR analysis which was prepared using the licensing basis computer code.Thus, it is concluded that the change in computer codes does not affect the results.As requested by this question, the suppression pool temperature and drywell and wetwell vapor space pressure profiles for this benchmark case are attached as Figures 1-1 and 1-2.
Given that the peak suppression pool temperature obtained from the SHEX UFSAR benchmark case is very close to the 177'limit for the BFN long-term torus integrity program (LTTIP), it is clear that with a direct incorporation of the power uprate input assumptions, the peak suppression pool temperature would exceed this containment design criteria.To resolve this issue, revised containment input parameters were imposed such that the net effect of power uprate will result in a peak suppression pool temperature of 177':~92'RHR service water temperature
~223 Btu/sec-'F
~4000 gpm RHR service water flow rate~6500 gpm RHR flow rate.The 92'RHR service water temperature is a BFN technical specification change item (included in the October 1, 1997 submittal as Surveillance Requirement 3.7.1.2 and associated Figure 3.7.1-1).Operating in accordance with proposed Figure 3.7.1-1 will allow'the plant to remain within its LTTIP limits (177')up to the current Ultimate Heat Sink limit of 95'.The RHR service water-flow rate is conservatively assumed at a lower value of 4000 gpm to more accurately reflect system performance with two RHRSW pumps supplying one loop of RHR heat exchangers.
The lower RHR heat exchanger k-factor is the result of the change to the RHR service water temperature and the RHR service water flow rate.Additionally, the ANSI/ANS 5.1-1979 plus 2a uncertainty decay heat curve was utilized as discussed in the letter from Ashok Thadani (NRC)to Gary Sozzi (GE),"Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.The results of this revised set of input parameters showed a peak suppression pool temperature of 175'and 177'for the pre-uprate and the power uprate condition, respectively.
The UFSAR will be revised to reflect these conditions upon Power Uprate implementation.
Please see the TVA Reply to NRC Request 2 for the uprated containment temperature response profiles.E-2 BRONNS FERRY (SP 7ENP 500.200.100.LLI LLI 0.2.LOG TINE-SEC 5.Figure 1-1: FSAR Benchmark Case-Suppression Pool Temperature Response at Pre-Uprate Power 60.BRONNS FERRY CONT RESPONSE TOL ASE C l AC DW PRESSURE IQ PRESSURE IO.(A CL 20.0.2.LOG TINE-SEC S.Figure 1-2: FSAR Benchmark Case-DryweH and Wetwell Pressure Response at Pre-Uprate Power NRC REQUEST 2 Section 4.1.1.2, Containment Gas Temperature Response indicates that at uprated power, the calculated peak drywell airspace temperature exceeds the drywell shell design temperature, but only at the beginning of the accident and for a short period of time.This is not considered a threat to the drywell shell structure due to relative long drywell shell heat-up time.Please provide the containment temperature profiles after a loss-of-coolant-accident.
TVA REPLY 2 The containment drywell and wetwell vapor space temperature profiles for Power Uprate conditions are provided in Figures 2-1 and 2-2.E-5 OASKLL TEtPERAT HE 150.500.U)LLI Ld CL CO LLI C3 UJ ISO.I-CL EU CL LU I-0.0.10.20.TINE (SECONOS)50.'!0.Figure 2-1: DBA-LOCA Containment Temperature Response at Power Uprate Condition l3ROHNS I ERRV CONf RESPONSE l9.ASE C SP fBP 2~LOG T IHE-SE.C Figure 2-2: DBA-LOCA Suppression Pool Temperature Response at Power Uprate Condition NRC REQUEST 3 Section 4.1.2.3, Subcompartment Pressurization, indicates that the sacrificial shield wall design remains adequate because the peak pressure in the annulus region increases slightly due to the power uprate.Please provide details of the results to justify your conclusion.
TVA REPLY 3 The structural evaluation in BFN UFSAR Section 12.2.2.6 demonstrates that the shield wall structure can withstand 19 psi pressurization, which is a differential pressure across the shield wall from the annulus to the drywell space.The largest line which has the safe-end located in the annulus is the 4-inch jet pump instrument line nozzle.For all larger lines, the double-ended line break results in the flow being directed into the drywell and not into the annulus.Effects of a postulated loss of coolant accident (LOCA)occurring within the sacrificial shield area have been investigated utilizing uprated conditions.
Method of Evaluation It is assumed that the design basis accident occurs and that the fluid is saturated through the break.For conservatism, Appendix K conditions (Moody's slip flow model)are also assumed.The following power/flow conditions will be used: Pre-Uprate Uprated Uprated (MELLLA)Dome Pressure'psia) 1023 1053 1053 Power (Mwt)3358 3527 3527 Flow (%Flow)100 100 81 Dome pressure increases 3 psi above nominal due to 2%power increase for Appendix K conditions.
Nominal conditions:
100%of rated power and 1020 psia dome pressure at pre-uprate conditions, and 100%of rated power and 1050 psia at uprated conditions.
where: 3293 MWt 3358 MWt 3458 MWt 3527 Mwt 102.5 Mlb/hr 100%of pze-uprate power 102%of pre-uprate power 100%of uprated power 102%of uprated power 100%of coze flow (pre-uprate
&uprate)The methods used are SAFER/GESTR-LOCA and engineering calculation.
E-8 The SAFER/GESTR-LOCA model is a thermal-hydraulic transient code developed by GE and approved by the NRC for long term inventory analysis of BWR LOCAs as well as other off-normal reactor transients where the neutron kinetics are of no consequence.
The SAFER/GESTR-LOCA model simulates all the BWR major vessel regions: lower plenum, guide tubes, core, core bypass, upper plenum, the initially subcooled region outside the core shroud, the saturated region outside the core shroud and the steam dome.In addition, flow, inventory, and heat transfer calculations are performed for a high power fuel assembly.The program solves the mass and energy balances for each region and the momentum equation for the flow loops.A single pressure is used for the thermodynamic property calculations.
The SAFER/GESTR-LOCA model can simulate all BWR water makeup systems.The pressure different:ial across the shield wall, hP, can be characterized based on the velocity in the annulus: AP=k(pV j where: k is the loss coefficient in the annulus p is t: he flow density V is the local velocity For conservatism, the break flow velocity Vb is equal to the local velocity V and the break flow density pb is equal to the flow density p.In terms of the free stream velocity, V, BP=kpV Vb b co By the continuity equation, m~V p,A, mb V b pbA~E-9
!
wher'e: ma Pa mb Ab is the annulus flow rate is the flow area between the annulus is the steam density in the annulus is the break flow rate is the break area is the break flow density and drywell Combining the equations above for be 2 2 Vb m, b,e=kpV-=kpb-Vp,A, 2 Pa a pbAb ma 1 k-Pb 2 mb 1 k-G Ab Pb where: G is the break flow mass flux k is a constant.The SAFER/GESTR-LOCA outputs were used to obtain the break mass flux and break enthalpy.The previous equation was used to ratio the pressure differential for the different power/flow conditions, 2 be=pe-p~2 Gz p For pre-uprate conditions, the pressure differential is less than 2 psi, as specified in BFN UFSAR, Section 12.2.2.6.The uprated analyses are based upon an assumed pre-uprated value of 2 psi.Summar of Results The comparison of the annulus pressurization loads at pre-uprate and uprate conditions are summarized as follows: Power/Flow
(%Power/%Flow)
Pre-uprate (102/100)Uprate (102/100)Uprated (102/81)Annulus Pressurization (psi)2.09 2.41 Increase Above Pre-uprate (bpsi)0.09 0.41 The results show that the annulus pressurization load on the shield wall is increased by approximately 0.4 psi due to the power uprate at the MELLLA point.The MELLLA point gives the largest annulus pressurization.
The results demonstrate that the E-10 annul.us p'ressurization loads at uprated conditions are still well below the limit of the BFN design basis value of 19 psi.NRC REQUEST 4 Section 4.1.4, Combustible Gas Control In Containment, predicts an earlier startup of the containment atmospheric dilution system at uprated power due to the increase in hydrogen and oxygen generation rates and that the predicted startup time will not result in system operation outside established design or operational restraints.
Please provide the time responses.
TVA REPLY 4 Containment Atmospheric Dilution (CAD)system operation is performed to maintain containment oxygen levels below 5%volume following a LOCA.The response of the containment and the CAD system following a LOCA is expected to be consistent with that shown in UFSAR Figures 5.2-14 and 15.These original, design basis evaluations indicate that CAD initiation will be required between 1 and 2 days after the LOCA based on the suppression chamber oxygen level.A 5%power uprate will increase the oxygen generation rate by approximately 5%since oxygen generation is a function of radiolysis which is approximately a direct function of core power.Based on a 5%increase in oxygen generation rate, CAD initiation will be required one to two hours earlier.This slight reduction in time is not critical to manual operator actions during a LOCA.REFERENCES'VA letter to NRC dated October 1, 1997, Browns Ferry Nuclear Plant (BFN)-Units 2 and 3-Technical Specification (TS)Change TS-384-Request For License Amendment for Power Uprate Operation 2.TVA letter to NRC dated March 16, 1998, Browns Ferry Nuclear Plant (BFN)-Units 2 and 3 Technical Specification (TS)No.384 Supplement 1-Request for License Amendment for Power Uprate Operation 3.TVA letter to NRC dated March 20, 1998, Browns Ferry.Nuclear Plant (BFN)-Units 2 and 3-Technical Specification (TS)Change TS-384-Request for License Amendment for Power Uprate Operation 4.NRC letter to TVA dated June 3, 1998, Browns Ferry Nuclear Plant, Units 2 and 3: Request for Additional Information Relating to Technical Specification Change No.TS-384-Power Uprate Operation (TAC Nos.M99711 and M99712)E-12